Letter Sequence Other |
|---|
|
|
MONTHYEARML20069K5311983-04-18018 April 1983
Forwards Response to Generic Ltr 83-10D Re Resolution of TMI Action Item II.K.3.5, Automatic Trip of Reactor Coolant Pumps, Reflecting Westinghouse Owners Group Plan for Resolution of Item
Project stage: Other
ML20198E3091985-10-10010 October 1985
Forwards Response to Generic Ltr 85-12 Re Implementation of TMI Action Item II.K.3.5, Automatic Trip of Reactor Coolant Pumps
Project stage: Other
ML20205N1091986-04-22022 April 1986
Forwards Interim Rept Re Util 851010 Response to Generic Ltr 85-12 Concerning Reactor Coolant Pump Trip Criteria.Util Should Arrange Telcon W/Reviewer Prior to Providing Formal Response & Shortly After Receipt of Ltr
Project stage: Other
ML20203F9501986-04-22022 April 1986
Clarifies Responses in Attachments 4 & 5 of Re Outstanding Issues 8(c) & 8(d).Attachment 5 Clarification Re Overloading of Emergency Diesel Generator Requires Corresponding Change to SER Section 8.3.1.15
Project stage: Request
ML20210H8281986-09-0505 September 1986
Forwards Supplemental Info to Util 830418 & 851010 Responses to Generic Ltrs 83-10d & 85-12 Re Reactor Coolant Pump Trip Issue,In Response to NRC 860422 Interim Rept
Project stage: Supplement
1985-10-10
[Table View] |
|
|---|
Category:CORRESPONDENCE-LETTERS
MONTHYEARIR 05000412/19990071999-10-21021 October 1999
Refers to Special Team Insp 50-412/99-07 Conducted from 990720-29 & Forwards Nov.Two Violations Identified.First Violation Involved Failure to Implement C/A to Prevent Biofouling of Service Water System
ML20217M1591999-10-19019 October 1999
Forwards NRC Rept Number 17, Requal Tracking Rept from Operator Licensing Tracking Sys.Rept Was Used by NRC to Schedule Requalification Exam for Operators & Record Requal Pass Dates
L-99-143, Informs That Subsequent Review of Approval Ltrs & SE for Relief Requests 1TYP-2-B5.40-1 (Rev 0),BV1-RV-AUG (Rev 1), BV2-RV-AUG (Rev 1),BV3-IWA-1 (Rev 1) & BV3-IWA-2 (Rev 1) Identified Erroneous Statements Which Dl Wishes to Correct1999-10-11011 October 1999
Informs That Subsequent Review of Approval Ltrs & SE for Relief Requests 1TYP-2-B5.40-1 (Rev 0),BV1-RV-AUG (Rev 1), BV2-RV-AUG (Rev 1),BV3-IWA-1 (Rev 1) & BV3-IWA-2 (Rev 1) Identified Erroneous Statements Which Dl Wishes to Correct
L-99-152, Submits Relief Request BV3-N-533-1,rev 0,requesting Use of ASME Approved Code Case N-553-1, Alternative Requirements for VT-2 Visual Examination of Class 1,2 & 3 Insulated Pressure Retaining Bolted Connections1999-10-11011 October 1999
Submits Relief Request BV3-N-533-1,rev 0,requesting Use of ASME Approved Code Case N-553-1, Alternative Requirements for VT-2 Visual Examination of Class 1,2 & 3 Insulated Pressure Retaining Bolted Connections
ML20217C6741999-10-0808 October 1999
Forwards RAI Re Licensee 970128 Response to GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions, . Response Requested within 60 Days of Receipt of Ltr
L-99-151, Responds to NRC Re Violations Noted in Insp Rept 50-412/99-07.Corrective Actions:Condition Repts Were Written for Listed Issues So That Repts Could Be Addressed Using BVPS Corrective Action Program1999-10-0707 October 1999
Responds to NRC Re Violations Noted in Insp Rept 50-412/99-07.Corrective Actions:Condition Repts Were Written for Listed Issues So That Repts Could Be Addressed Using BVPS Corrective Action Program
ML20217E0301999-10-0707 October 1999
Forwards Insp Repts 50-334/99-06 & 50-412/99-06 on 990809-13 & 990823-27.Violation Noted Involving Failure to Correctly Translate Design Change Re Pertinent Operating Logs & Plant Equipment Labeling
ML20212M2661999-09-30030 September 1999
Forwards Order Approving Transfer of Licenses for Beaver Valley from Dlc to Pennsylvania Power Co & Approving Conforming Amends in Response to 990505 Application
ML20212K8071999-09-30030 September 1999
Informs That on 990916,NRC Staff Completed mid-cycle Plant Performance Review (PPR) of Facility.Staff Conducted Reviews of All Operating NPPs to Integrate Performance Info & to Plan for Insp Activities at Facility
ML20216J9621999-09-30030 September 1999
Forwards Insp Repts 50-334/99-05 & 50-412/99-05 on 990725-0904.Two Violations Noted & Being Treated as Ncvs.One Violation Re Failure to Follow Operation Manual Procedure Associated with Configuration Control Identified
L-99-149, Informs NRC That Items Identified in 990629 Response to GL 98-01 Have Been Completed.Attached Table Submitted with Has Been Updated to Reflect Completion1999-09-28028 September 1999
Informs NRC That Items Identified in 990629 Response to GL 98-01 Have Been Completed.Attached Table Submitted with Has Been Updated to Reflect Completion
L-99-148, Notifies NRC of License Withdrawal of M Linch,License SOP-11478 IAW 10CFR50.74.M Linch Resigned from Employment at Bvps,Effective 9908171999-09-24024 September 1999
Notifies NRC of License Withdrawal of M Linch,License SOP-11478 IAW 10CFR50.74.M Linch Resigned from Employment at Bvps,Effective 990817
ML20212G0601999-09-23023 September 1999
Forwards Answer of Duquesne Light Co to Petition to Waive Time Limits & Suppl Comments of Local 29, Intl Brotherhood of Electrical Workers.Copies of Answer Have Been Served to Parties & Petitioner by e-mail or Facsimile
ML20212C5521999-09-21021 September 1999
Forwards for Filing,Answer to Firstenergy Nuclear Operating Co & Pennsylvania Power Co in Opposition to Petition to Waive Time Limits & Suppl Comments of Local 29 Intl Brotherhood of Electrical Workers
L-99-144, Forwards NRC Form 536 Which Addresses Util Proposed Operator Licensing Exam Schedule,In Response to Administrative Ltr 99-031999-09-20020 September 1999
Forwards NRC Form 536 Which Addresses Util Proposed Operator Licensing Exam Schedule,In Response to Administrative Ltr 99-03
ML20212B3291999-09-16016 September 1999
Forwards for Filing,Petition to Waive Time Limits in 10CFR2.1305 & Supplemental Comments of Local 29,Intl Brotherhood of Electrical Workers Re Beaver Valley Power Station,Units 1 & 2
L-99-134, Provides Addl Info to Support 990617 LAR 127,proposing Mods to Heatup,Cooldown & Overpressure Protection Curves.Info Is Provided to Formally Docket Info Discussed During 990820 Telcon with NRC1999-09-15015 September 1999
Provides Addl Info to Support 990617 LAR 127,proposing Mods to Heatup,Cooldown & Overpressure Protection Curves.Info Is Provided to Formally Docket Info Discussed During 990820 Telcon with NRC
ML20211Q3431999-09-0808 September 1999
Informs That During 990903 Telcon Between L Briggs & T Kuhar,Arrangements Were Made for NRC to Inspect Licensed Operator Requalification Program at Plant,Unit 1.Insp Planned for Wk of 991115
ML20211Q5601999-09-0707 September 1999
Forwards Insp Rept 50-412/99-07 on 990720-29.Three Apparent Violations Noted & Being Considered for Escalated Ea. Violations Involve Failure to Implement C/As to Prevent bio- Fouling of Svc Water Sys
L-99-138, Forwards Rev 37 to Issue 4 for BVPS Physical Security Plan (PSP) (Base Plan,Safeguards Contingency Plan & Training & Qualification Plan).Changes to Plan Are Listed.Encls Withheld Per 10CFR2.790(d)1999-09-0303 September 1999
Forwards Rev 37 to Issue 4 for BVPS Physical Security Plan (PSP) (Base Plan,Safeguards Contingency Plan & Training & Qualification Plan).Changes to Plan Are Listed.Encls Withheld Per 10CFR2.790(d)
L-99-136, Forwards Data Point Library (Dpl) Changes,Iaw Provisions of 10CFR50,App E,Section VI.3.a & NUREG-1394,Rev 1,Section 3.6, Administrative Implementation Requirements. Ten Listed Dpls Have Specific Setpoint Values Removed,Per Unit 2 Dpls1999-09-0202 September 1999
Forwards Data Point Library (Dpl) Changes,Iaw Provisions of 10CFR50,App E,Section VI.3.a & NUREG-1394,Rev 1,Section 3.6, Administrative Implementation Requirements. Ten Listed Dpls Have Specific Setpoint Values Removed,Per Unit 2 Dpls
L-99-098, Forwards Proposed Changes to Bvps,Units 1 & 2 Operations QA Program Description,Iaw 10CFR50.54(a)(3)(ii).Change Would Reduce Operations QA Program Description Commitments by Limiting Required Onsite Safety Committee Reviews of Mods1999-09-0202 September 1999
Forwards Proposed Changes to Bvps,Units 1 & 2 Operations QA Program Description,Iaw 10CFR50.54(a)(3)(ii).Change Would Reduce Operations QA Program Description Commitments by Limiting Required Onsite Safety Committee Reviews of Mods
L-99-137, Forwards Issue 3,rev 3 to Bvps,Unit 1 Inservice Testing Program for Pumps & Valves. Encl 1 Provides Summary of IST Program Changes Which Have Been Incorporated Into Issue 3, Rev 31999-08-31031 August 1999
Forwards Issue 3,rev 3 to Bvps,Unit 1 Inservice Testing Program for Pumps & Valves. Encl 1 Provides Summary of IST Program Changes Which Have Been Incorporated Into Issue 3, Rev 3
L-99-022, Forwards Issue 2,Rev 1 to Bvps,Unit 2 Inservice Testing (IST) Program for Pumps & Valves. Summary of Changes,Encl1999-08-31031 August 1999
Forwards Issue 2,Rev 1 to Bvps,Unit 2 Inservice Testing (IST) Program for Pumps & Valves. Summary of Changes,Encl
L-99-012, Forwards Proposed Change to Bvps,Units 1 & 2,operations QA Program Description,Per 10CFR50.54(a)(3)(ii),including Description of Proposed Change,Reason for Change & Basis for Concluding Revised Program Satisfy 10CFR50,App B1999-08-30030 August 1999
Forwards Proposed Change to Bvps,Units 1 & 2,operations QA Program Description,Per 10CFR50.54(a)(3)(ii),including Description of Proposed Change,Reason for Change & Basis for Concluding Revised Program Satisfy 10CFR50,App B
L-99-037, Forwards Proposed Change to Bvps,Units 1 & 2 Operations QA Program Description,Iaw 10CFR50.54(a)(3)(ii).Attachment 1 Further Describes Proposed Change & Identifies Reason for Change1999-08-30030 August 1999
Forwards Proposed Change to Bvps,Units 1 & 2 Operations QA Program Description,Iaw 10CFR50.54(a)(3)(ii).Attachment 1 Further Describes Proposed Change & Identifies Reason for Change
L-99-132, Forwards fitness-for-duty Program Six Month Rept for 990101-990630 for Bvps,Units 1 & 21999-08-26026 August 1999
Forwards fitness-for-duty Program Six Month Rept for 990101-990630 for Bvps,Units 1 & 2
05000412/LER-1999-007, Forwards LER 99-007-00, Forced Shutdown Due to Inoperable EDG, Per 10CFR50.73(a)(2)(i).Rept Is Delayed Due to Util Needing Addl Three Days to Address Event Issues on Reportability & Provide Addl Safety Implications Info1999-08-19019 August 1999
Forwards LER 99-007-00, Forced Shutdown Due to Inoperable EDG, Per 10CFR50.73(a)(2)(i).Rept Is Delayed Due to Util Needing Addl Three Days to Address Event Issues on Reportability & Provide Addl Safety Implications Info
ML20211A5111999-08-18018 August 1999
Forwards Insp Repts 50-334/99-04 & 50-412/99-04 on 990613- 990724.One Violation Noted & Treated as Non-Cited Violation Involved Failure to Maintain Containment Equipment Hatch Closed During Fuel Movement
L-99-127, Provides Response to NRC Ltr Requesting Review & Comment of NRC Reactor Vessel Structural Integrity Database by 990901.Inconsistencies Noted in Way Data Characterizes Condition of Reactor Vessel1999-08-17017 August 1999
Provides Response to NRC Ltr Requesting Review & Comment of NRC Reactor Vessel Structural Integrity Database by 990901.Inconsistencies Noted in Way Data Characterizes Condition of Reactor Vessel
L-99-124, Requests Withdrawal of Editorial Changes That Do Not Pertain to Transfer of Operating Authority or Plant Ownership That Are Proposed in LARs 269 & 144.Revised mark-up License Pages Reflecting Changes,Attached1999-07-30030 July 1999
Requests Withdrawal of Editorial Changes That Do Not Pertain to Transfer of Operating Authority or Plant Ownership That Are Proposed in LARs 269 & 144.Revised mark-up License Pages Reflecting Changes,Attached
L-99-121, Submits Data Point Library (Dpl) Changes,Iaw Requirements of 10CFR50,App E,Section VI.3.a & NUREG-1394,rev 1,Section 3.6, Administrative Implementation Reqiurements1999-07-28028 July 1999
Submits Data Point Library (Dpl) Changes,Iaw Requirements of 10CFR50,App E,Section VI.3.a & NUREG-1394,rev 1,Section 3.6, Administrative Implementation Reqiurements
L-99-118, Forwards Response to NRC AL 99-02,request for Info Re Estimate of Number of New Licensing Actions Expected to Be Submitted in Fy 2000 & 20011999-07-25025 July 1999
Forwards Response to NRC AL 99-02,request for Info Re Estimate of Number of New Licensing Actions Expected to Be Submitted in Fy 2000 & 2001
L-99-120, Forwards Annual Rept of Facility Changes,Tests & Experiments for BVPS Unit 1,IAW 10CFR50.59.Rept Provides Brief Description of Each Facility & Procedure Change & Summary of Safety Evaluations1999-07-22022 July 1999
Forwards Annual Rept of Facility Changes,Tests & Experiments for BVPS Unit 1,IAW 10CFR50.59.Rept Provides Brief Description of Each Facility & Procedure Change & Summary of Safety Evaluations
L-99-119, Forwards Rev 17 to UFSAR for Beaver Power Station,Unit 1. Submittal Reflects Changes to Facility & Procedures as Described in UFSAR That Were Completed During Annual Reporting Period Endinig 9901221999-07-20020 July 1999
Forwards Rev 17 to UFSAR for Beaver Power Station,Unit 1. Submittal Reflects Changes to Facility & Procedures as Described in UFSAR That Were Completed During Annual Reporting Period Endinig 990122
L-99-113, Forwards Final,Typed TS Pages for LARs 262 & 135,previously Submitted with Editorial Changes Identified by1999-07-15015 July 1999
Forwards Final,Typed TS Pages for LARs 262 & 135,previously Submitted with Editorial Changes Identified by
L-99-111, Forwards Revised Final Typed Pages for LARs 109 & 115, Previously Submitted by 990615 & 28 Ltrs.Revised Pages Replace Those Previously Provided for LARs 109 & 115.With Summary of Editorial Changes1999-07-15015 July 1999
Forwards Revised Final Typed Pages for LARs 109 & 115, Previously Submitted by 990615 & 28 Ltrs.Revised Pages Replace Those Previously Provided for LARs 109 & 115.With Summary of Editorial Changes
L-99-112, Responds to NRC Questions Made During 990518 Meeting with Util Re LARs 220 & 88,for Bvps,Units 1 & 2.Copy of Ltr DLC-99-743,which Is non-proprietary Version of DLC-96-310 & Westinghouse Technical Bulletin ESBU-TB-96-07-R0 Also Encl1999-07-14014 July 1999
Responds to NRC Questions Made During 990518 Meeting with Util Re LARs 220 & 88,for Bvps,Units 1 & 2.Copy of Ltr DLC-99-743,which Is non-proprietary Version of DLC-96-310 & Westinghouse Technical Bulletin ESBU-TB-96-07-R0 Also Encl
L-99-110, Forwards Changes to ERDS for Unit 1,IAW Requirements of 10CFR50,App E,Section VI.3.a & NUREG-1394,Rev 1,Section 3.61999-07-14014 July 1999
Forwards Changes to ERDS for Unit 1,IAW Requirements of 10CFR50,App E,Section VI.3.a & NUREG-1394,Rev 1,Section 3.6
ML20209G5701999-07-12012 July 1999
Discusses Closure of TACs MA0525 & MA0526 Re Response to RAI Concerning GL 92-0,Rev 1,Suppl 1, Rv Structural Integrity. Info in Rvid Revised & Released as Ver 2 as Result of Review of Response
ML20207H6621999-07-0808 July 1999
Forwards RAI Re Util 981112 Response to IPEEE Evaluations for Plant,Units 1 & 2.RAI Was Discussed During 990628 Telcon in Order to Ensure Clear Consistent Understanding by All Parties of Info Needed
L-99-105, Forwards Response to NRC 990420 RAI Re Previous Responses to GL 95-07, Pressure Locking & Thermal Binding of SR Power-Operated Gate Valves1999-07-0808 July 1999
Forwards Response to NRC 990420 RAI Re Previous Responses to GL 95-07, Pressure Locking & Thermal Binding of SR Power-Operated Gate Valves
ML20209D8191999-07-0707 July 1999
Forwards Insp Repts 50-334/99-03 & 50-412/99-03 on 990502- 0612.No Violations Noted.Program for Maintaining Occupational Exposures as Low as Reasonably Achievable (ALARA) & for Training Personnel,Generally Effective
L-99-109, Forwards Inservice Insp Ninety-Day Rept Bvps,Unit 2 Outage 7,Year 1999, IAW ASME Boiler & Pressure Vessel Code,Section Xi,Article IWA-62301999-07-0707 July 1999
Forwards Inservice Insp Ninety-Day Rept Bvps,Unit 2 Outage 7,Year 1999, IAW ASME Boiler & Pressure Vessel Code,Section Xi,Article IWA-6230
L-99-108, Requests Withdrawal of Change Proposed for TS Bases Page B 3/4 2-2 from LARs 1A-262 & 2A-135,originally Submitted by Licensee to NRC1999-07-0707 July 1999
Requests Withdrawal of Change Proposed for TS Bases Page B 3/4 2-2 from LARs 1A-262 & 2A-135,originally Submitted by Licensee to NRC
L-99-104, Responds to GL 98-01,Suppl 1, Y2K Readiness at Nuclear Power Plants. Disclosure Rept Encl1999-06-29029 June 1999
Responds to GL 98-01,Suppl 1, Y2K Readiness at Nuclear Power Plants. Disclosure Rept Encl
L-99-093, Provides Response to RAI on Proposed Change to Operations QA Program Description.Attachment 2 Provides Revised Markup for Proposed Changes to Affected UFSAR Section 17.2.51999-06-25025 June 1999
Provides Response to RAI on Proposed Change to Operations QA Program Description.Attachment 2 Provides Revised Markup for Proposed Changes to Affected UFSAR Section 17.2.5
L-99-102, Forwards Typed,Final TS Pages for LARs 259 & 131.Summary of Description of Plant Editorial Changes Incorporated in Addition to Those Addressed in Previously Provided marked-up Pages,Encl1999-06-22022 June 1999
Forwards Typed,Final TS Pages for LARs 259 & 131.Summary of Description of Plant Editorial Changes Incorporated in Addition to Those Addressed in Previously Provided marked-up Pages,Encl
L-99-101, Submits Response to NRC Oral RAI Concerning Qualifications for Senior Nuclear Executive Ref in 10CFR50.80 Submittal1999-06-22022 June 1999
Submits Response to NRC Oral RAI Concerning Qualifications for Senior Nuclear Executive Ref in 10CFR50.80 Submittal
L-99-062, Forwards Final TS Pages for LARs 262 & 135,including Summary Description of Plants Editorial Changes Incorporated in Addition to Those Addressed in Previously Provided marked-up Pages1999-06-17017 June 1999
Forwards Final TS Pages for LARs 262 & 135,including Summary Description of Plants Editorial Changes Incorporated in Addition to Those Addressed in Previously Provided marked-up Pages
1999-09-08
[Table view]
Category:INCOMING CORRESPONDENCE
MONTHYEARL-99-152, Submits Relief Request BV3-N-533-1,rev 0,requesting Use of ASME Approved Code Case N-553-1, Alternative Requirements for VT-2 Visual Examination of Class 1,2 & 3 Insulated Pressure Retaining Bolted Connections1999-10-11011 October 1999
Submits Relief Request BV3-N-533-1,rev 0,requesting Use of ASME Approved Code Case N-553-1, Alternative Requirements for VT-2 Visual Examination of Class 1,2 & 3 Insulated Pressure Retaining Bolted Connections
L-99-143, Informs That Subsequent Review of Approval Ltrs & SE for Relief Requests 1TYP-2-B5.40-1 (Rev 0),BV1-RV-AUG (Rev 1), BV2-RV-AUG (Rev 1),BV3-IWA-1 (Rev 1) & BV3-IWA-2 (Rev 1) Identified Erroneous Statements Which Dl Wishes to Correct1999-10-11011 October 1999
Informs That Subsequent Review of Approval Ltrs & SE for Relief Requests 1TYP-2-B5.40-1 (Rev 0),BV1-RV-AUG (Rev 1), BV2-RV-AUG (Rev 1),BV3-IWA-1 (Rev 1) & BV3-IWA-2 (Rev 1) Identified Erroneous Statements Which Dl Wishes to Correct
L-99-151, Responds to NRC Re Violations Noted in Insp Rept 50-412/99-07.Corrective Actions:Condition Repts Were Written for Listed Issues So That Repts Could Be Addressed Using BVPS Corrective Action Program1999-10-0707 October 1999
Responds to NRC Re Violations Noted in Insp Rept 50-412/99-07.Corrective Actions:Condition Repts Were Written for Listed Issues So That Repts Could Be Addressed Using BVPS Corrective Action Program
L-99-149, Informs NRC That Items Identified in 990629 Response to GL 98-01 Have Been Completed.Attached Table Submitted with Has Been Updated to Reflect Completion1999-09-28028 September 1999
Informs NRC That Items Identified in 990629 Response to GL 98-01 Have Been Completed.Attached Table Submitted with Has Been Updated to Reflect Completion
L-99-148, Notifies NRC of License Withdrawal of M Linch,License SOP-11478 IAW 10CFR50.74.M Linch Resigned from Employment at Bvps,Effective 9908171999-09-24024 September 1999
Notifies NRC of License Withdrawal of M Linch,License SOP-11478 IAW 10CFR50.74.M Linch Resigned from Employment at Bvps,Effective 990817
ML20212G0601999-09-23023 September 1999
Forwards Answer of Duquesne Light Co to Petition to Waive Time Limits & Suppl Comments of Local 29, Intl Brotherhood of Electrical Workers.Copies of Answer Have Been Served to Parties & Petitioner by e-mail or Facsimile
ML20212C5521999-09-21021 September 1999
Forwards for Filing,Answer to Firstenergy Nuclear Operating Co & Pennsylvania Power Co in Opposition to Petition to Waive Time Limits & Suppl Comments of Local 29 Intl Brotherhood of Electrical Workers
L-99-144, Forwards NRC Form 536 Which Addresses Util Proposed Operator Licensing Exam Schedule,In Response to Administrative Ltr 99-031999-09-20020 September 1999
Forwards NRC Form 536 Which Addresses Util Proposed Operator Licensing Exam Schedule,In Response to Administrative Ltr 99-03
ML20212B3291999-09-16016 September 1999
Forwards for Filing,Petition to Waive Time Limits in 10CFR2.1305 & Supplemental Comments of Local 29,Intl Brotherhood of Electrical Workers Re Beaver Valley Power Station,Units 1 & 2
L-99-134, Provides Addl Info to Support 990617 LAR 127,proposing Mods to Heatup,Cooldown & Overpressure Protection Curves.Info Is Provided to Formally Docket Info Discussed During 990820 Telcon with NRC1999-09-15015 September 1999
Provides Addl Info to Support 990617 LAR 127,proposing Mods to Heatup,Cooldown & Overpressure Protection Curves.Info Is Provided to Formally Docket Info Discussed During 990820 Telcon with NRC
L-99-138, Forwards Rev 37 to Issue 4 for BVPS Physical Security Plan (PSP) (Base Plan,Safeguards Contingency Plan & Training & Qualification Plan).Changes to Plan Are Listed.Encls Withheld Per 10CFR2.790(d)1999-09-0303 September 1999
Forwards Rev 37 to Issue 4 for BVPS Physical Security Plan (PSP) (Base Plan,Safeguards Contingency Plan & Training & Qualification Plan).Changes to Plan Are Listed.Encls Withheld Per 10CFR2.790(d)
L-99-136, Forwards Data Point Library (Dpl) Changes,Iaw Provisions of 10CFR50,App E,Section VI.3.a & NUREG-1394,Rev 1,Section 3.6, Administrative Implementation Requirements. Ten Listed Dpls Have Specific Setpoint Values Removed,Per Unit 2 Dpls1999-09-0202 September 1999
Forwards Data Point Library (Dpl) Changes,Iaw Provisions of 10CFR50,App E,Section VI.3.a & NUREG-1394,Rev 1,Section 3.6, Administrative Implementation Requirements. Ten Listed Dpls Have Specific Setpoint Values Removed,Per Unit 2 Dpls
L-99-098, Forwards Proposed Changes to Bvps,Units 1 & 2 Operations QA Program Description,Iaw 10CFR50.54(a)(3)(ii).Change Would Reduce Operations QA Program Description Commitments by Limiting Required Onsite Safety Committee Reviews of Mods1999-09-0202 September 1999
Forwards Proposed Changes to Bvps,Units 1 & 2 Operations QA Program Description,Iaw 10CFR50.54(a)(3)(ii).Change Would Reduce Operations QA Program Description Commitments by Limiting Required Onsite Safety Committee Reviews of Mods
L-99-137, Forwards Issue 3,rev 3 to Bvps,Unit 1 Inservice Testing Program for Pumps & Valves. Encl 1 Provides Summary of IST Program Changes Which Have Been Incorporated Into Issue 3, Rev 31999-08-31031 August 1999
Forwards Issue 3,rev 3 to Bvps,Unit 1 Inservice Testing Program for Pumps & Valves. Encl 1 Provides Summary of IST Program Changes Which Have Been Incorporated Into Issue 3, Rev 3
L-99-022, Forwards Issue 2,Rev 1 to Bvps,Unit 2 Inservice Testing (IST) Program for Pumps & Valves. Summary of Changes,Encl1999-08-31031 August 1999
Forwards Issue 2,Rev 1 to Bvps,Unit 2 Inservice Testing (IST) Program for Pumps & Valves. Summary of Changes,Encl
L-99-037, Forwards Proposed Change to Bvps,Units 1 & 2 Operations QA Program Description,Iaw 10CFR50.54(a)(3)(ii).Attachment 1 Further Describes Proposed Change & Identifies Reason for Change1999-08-30030 August 1999
Forwards Proposed Change to Bvps,Units 1 & 2 Operations QA Program Description,Iaw 10CFR50.54(a)(3)(ii).Attachment 1 Further Describes Proposed Change & Identifies Reason for Change
L-99-012, Forwards Proposed Change to Bvps,Units 1 & 2,operations QA Program Description,Per 10CFR50.54(a)(3)(ii),including Description of Proposed Change,Reason for Change & Basis for Concluding Revised Program Satisfy 10CFR50,App B1999-08-30030 August 1999
Forwards Proposed Change to Bvps,Units 1 & 2,operations QA Program Description,Per 10CFR50.54(a)(3)(ii),including Description of Proposed Change,Reason for Change & Basis for Concluding Revised Program Satisfy 10CFR50,App B
L-99-132, Forwards fitness-for-duty Program Six Month Rept for 990101-990630 for Bvps,Units 1 & 21999-08-26026 August 1999
Forwards fitness-for-duty Program Six Month Rept for 990101-990630 for Bvps,Units 1 & 2
05000412/LER-1999-007, Forwards LER 99-007-00, Forced Shutdown Due to Inoperable EDG, Per 10CFR50.73(a)(2)(i).Rept Is Delayed Due to Util Needing Addl Three Days to Address Event Issues on Reportability & Provide Addl Safety Implications Info1999-08-19019 August 1999
Forwards LER 99-007-00, Forced Shutdown Due to Inoperable EDG, Per 10CFR50.73(a)(2)(i).Rept Is Delayed Due to Util Needing Addl Three Days to Address Event Issues on Reportability & Provide Addl Safety Implications Info
L-99-127, Provides Response to NRC Ltr Requesting Review & Comment of NRC Reactor Vessel Structural Integrity Database by 990901.Inconsistencies Noted in Way Data Characterizes Condition of Reactor Vessel1999-08-17017 August 1999
Provides Response to NRC Ltr Requesting Review & Comment of NRC Reactor Vessel Structural Integrity Database by 990901.Inconsistencies Noted in Way Data Characterizes Condition of Reactor Vessel
L-99-124, Requests Withdrawal of Editorial Changes That Do Not Pertain to Transfer of Operating Authority or Plant Ownership That Are Proposed in LARs 269 & 144.Revised mark-up License Pages Reflecting Changes,Attached1999-07-30030 July 1999
Requests Withdrawal of Editorial Changes That Do Not Pertain to Transfer of Operating Authority or Plant Ownership That Are Proposed in LARs 269 & 144.Revised mark-up License Pages Reflecting Changes,Attached
L-99-121, Submits Data Point Library (Dpl) Changes,Iaw Requirements of 10CFR50,App E,Section VI.3.a & NUREG-1394,rev 1,Section 3.6, Administrative Implementation Reqiurements1999-07-28028 July 1999
Submits Data Point Library (Dpl) Changes,Iaw Requirements of 10CFR50,App E,Section VI.3.a & NUREG-1394,rev 1,Section 3.6, Administrative Implementation Reqiurements
L-99-118, Forwards Response to NRC AL 99-02,request for Info Re Estimate of Number of New Licensing Actions Expected to Be Submitted in Fy 2000 & 20011999-07-25025 July 1999
Forwards Response to NRC AL 99-02,request for Info Re Estimate of Number of New Licensing Actions Expected to Be Submitted in Fy 2000 & 2001
L-99-120, Forwards Annual Rept of Facility Changes,Tests & Experiments for BVPS Unit 1,IAW 10CFR50.59.Rept Provides Brief Description of Each Facility & Procedure Change & Summary of Safety Evaluations1999-07-22022 July 1999
Forwards Annual Rept of Facility Changes,Tests & Experiments for BVPS Unit 1,IAW 10CFR50.59.Rept Provides Brief Description of Each Facility & Procedure Change & Summary of Safety Evaluations
L-99-119, Forwards Rev 17 to UFSAR for Beaver Power Station,Unit 1. Submittal Reflects Changes to Facility & Procedures as Described in UFSAR That Were Completed During Annual Reporting Period Endinig 9901221999-07-20020 July 1999
Forwards Rev 17 to UFSAR for Beaver Power Station,Unit 1. Submittal Reflects Changes to Facility & Procedures as Described in UFSAR That Were Completed During Annual Reporting Period Endinig 990122
L-99-111, Forwards Revised Final Typed Pages for LARs 109 & 115, Previously Submitted by 990615 & 28 Ltrs.Revised Pages Replace Those Previously Provided for LARs 109 & 115.With Summary of Editorial Changes1999-07-15015 July 1999
Forwards Revised Final Typed Pages for LARs 109 & 115, Previously Submitted by 990615 & 28 Ltrs.Revised Pages Replace Those Previously Provided for LARs 109 & 115.With Summary of Editorial Changes
L-99-113, Forwards Final,Typed TS Pages for LARs 262 & 135,previously Submitted with Editorial Changes Identified by1999-07-15015 July 1999
Forwards Final,Typed TS Pages for LARs 262 & 135,previously Submitted with Editorial Changes Identified by
L-99-112, Responds to NRC Questions Made During 990518 Meeting with Util Re LARs 220 & 88,for Bvps,Units 1 & 2.Copy of Ltr DLC-99-743,which Is non-proprietary Version of DLC-96-310 & Westinghouse Technical Bulletin ESBU-TB-96-07-R0 Also Encl1999-07-14014 July 1999
Responds to NRC Questions Made During 990518 Meeting with Util Re LARs 220 & 88,for Bvps,Units 1 & 2.Copy of Ltr DLC-99-743,which Is non-proprietary Version of DLC-96-310 & Westinghouse Technical Bulletin ESBU-TB-96-07-R0 Also Encl
L-99-110, Forwards Changes to ERDS for Unit 1,IAW Requirements of 10CFR50,App E,Section VI.3.a & NUREG-1394,Rev 1,Section 3.61999-07-14014 July 1999
Forwards Changes to ERDS for Unit 1,IAW Requirements of 10CFR50,App E,Section VI.3.a & NUREG-1394,Rev 1,Section 3.6
L-99-105, Forwards Response to NRC 990420 RAI Re Previous Responses to GL 95-07, Pressure Locking & Thermal Binding of SR Power-Operated Gate Valves1999-07-0808 July 1999
Forwards Response to NRC 990420 RAI Re Previous Responses to GL 95-07, Pressure Locking & Thermal Binding of SR Power-Operated Gate Valves
L-99-108, Requests Withdrawal of Change Proposed for TS Bases Page B 3/4 2-2 from LARs 1A-262 & 2A-135,originally Submitted by Licensee to NRC1999-07-0707 July 1999
Requests Withdrawal of Change Proposed for TS Bases Page B 3/4 2-2 from LARs 1A-262 & 2A-135,originally Submitted by Licensee to NRC
L-99-109, Forwards Inservice Insp Ninety-Day Rept Bvps,Unit 2 Outage 7,Year 1999, IAW ASME Boiler & Pressure Vessel Code,Section Xi,Article IWA-62301999-07-0707 July 1999
Forwards Inservice Insp Ninety-Day Rept Bvps,Unit 2 Outage 7,Year 1999, IAW ASME Boiler & Pressure Vessel Code,Section Xi,Article IWA-6230
L-99-104, Responds to GL 98-01,Suppl 1, Y2K Readiness at Nuclear Power Plants. Disclosure Rept Encl1999-06-29029 June 1999
Responds to GL 98-01,Suppl 1, Y2K Readiness at Nuclear Power Plants. Disclosure Rept Encl
L-99-093, Provides Response to RAI on Proposed Change to Operations QA Program Description.Attachment 2 Provides Revised Markup for Proposed Changes to Affected UFSAR Section 17.2.51999-06-25025 June 1999
Provides Response to RAI on Proposed Change to Operations QA Program Description.Attachment 2 Provides Revised Markup for Proposed Changes to Affected UFSAR Section 17.2.5
L-99-102, Forwards Typed,Final TS Pages for LARs 259 & 131.Summary of Description of Plant Editorial Changes Incorporated in Addition to Those Addressed in Previously Provided marked-up Pages,Encl1999-06-22022 June 1999
Forwards Typed,Final TS Pages for LARs 259 & 131.Summary of Description of Plant Editorial Changes Incorporated in Addition to Those Addressed in Previously Provided marked-up Pages,Encl
L-99-101, Submits Response to NRC Oral RAI Concerning Qualifications for Senior Nuclear Executive Ref in 10CFR50.80 Submittal1999-06-22022 June 1999
Submits Response to NRC Oral RAI Concerning Qualifications for Senior Nuclear Executive Ref in 10CFR50.80 Submittal
L-99-062, Forwards Final TS Pages for LARs 262 & 135,including Summary Description of Plants Editorial Changes Incorporated in Addition to Those Addressed in Previously Provided marked-up Pages1999-06-17017 June 1999
Forwards Final TS Pages for LARs 262 & 135,including Summary Description of Plants Editorial Changes Incorporated in Addition to Those Addressed in Previously Provided marked-up Pages
ML20196A6601999-06-16016 June 1999
Forwards Master Decommissioning Trust Agreements Revised After 1990 for Ohio Edison Co,Cleveland Electric Illuminating Co,Toledo Edison Co & Pennsylvania Power Co Re Bvnps,Units 1 & 2,DBNPS,Unit 1 & Perry Unit 1
ML20195H4651999-06-16016 June 1999
Forwards for Filing Answer of Firstenergy Corp in Opposition to Petition for Leave to Intervene of Local 29, Intl Brotherhood of Electrical Workers. Copies of Answer Have Been Served Upon Parties & Petitioner by e-mail
ML20195J5221999-06-16016 June 1999
Forwards Answer of Duquesne Light Co to Petition to Intervene of Local 29,International Brotherhood of Electrical Workers in Listed Matter.With Certificate of Svc
L-99-100, Forwards Typed,Final TS Pages for LAR 109 Re Rcs.Summary Description of Editorial Changes Incorporated in Addition to Those Addressed in Previously Provided marked-up Pages Is Provided in Attachment B-1091999-06-15015 June 1999
Forwards Typed,Final TS Pages for LAR 109 Re Rcs.Summary Description of Editorial Changes Incorporated in Addition to Those Addressed in Previously Provided marked-up Pages Is Provided in Attachment B-109
L-99-095, Provides Addl Info to Support LARs 262 & 135,in Response to NRC 990527 Verbal Request.Info Describes Performance of Unit 1 Rod Position Indication Sys & Provides Some Background Info Re Normal Operation of Sys1999-06-15015 June 1999
Provides Addl Info to Support LARs 262 & 135,in Response to NRC 990527 Verbal Request.Info Describes Performance of Unit 1 Rod Position Indication Sys & Provides Some Background Info Re Normal Operation of Sys
L-99-099, Requests Partial Withdrawal of LAR 120,which Requested Review of USQ Due to Increased Calculated Doses for Locked Rotor Event & Use of Unapproved Methodology for Evaluating Small Break LOCA Doses Involving W Natl Safety Advisory Ltr1999-06-14014 June 1999
Requests Partial Withdrawal of LAR 120,which Requested Review of USQ Due to Increased Calculated Doses for Locked Rotor Event & Use of Unapproved Methodology for Evaluating Small Break LOCA Doses Involving W Natl Safety Advisory Ltr
ML20195H3731999-06-0303 June 1999
Forwards Petition to Intervene of Local 29,Intl Brotherhood of Electrical Workers in Matter of Firstenergy Nuclear Operating Co,For Filing
L-99-090, Forwards Summary Review Completed to Verify Adequacy of Design Basis Accident Thermal Overpressure Protection for BVPS Unit 2 Containment Penetrations,Per Request1999-06-0202 June 1999
Forwards Summary Review Completed to Verify Adequacy of Design Basis Accident Thermal Overpressure Protection for BVPS Unit 2 Containment Penetrations,Per Request
L-99-086, Forwards Bvps,Unit 2 SG Exam Rept for Aug 1998.Rept Provided to Document Results of SG Eddy Current Exams Performed in Aug 1998.Summary of Insps Provided in Encl 11999-05-28028 May 1999
Forwards Bvps,Unit 2 SG Exam Rept for Aug 1998.Rept Provided to Document Results of SG Eddy Current Exams Performed in Aug 1998.Summary of Insps Provided in Encl 1
L-99-089, Forwards Annual Financial Repts,Including Certified Financial Statements,Of Dqe,Firstenergy Corp,Ohio Edison Co,Pennsylvania Power Co,Cleveland Electric Illuminating Co & Toledo Edison Co,Iaw 10CFR50.71(b)1999-05-28028 May 1999
Forwards Annual Financial Repts,Including Certified Financial Statements,Of Dqe,Firstenergy Corp,Ohio Edison Co,Pennsylvania Power Co,Cleveland Electric Illuminating Co & Toledo Edison Co,Iaw 10CFR50.71(b)
L-99-084, Forwards Revised marked-up TS & UFSAR Pages to 990303 LARs 259 & 131 Which Revised Qualifications for Operations Mgt & Incorporated Generic Position Titles in Ts.Encl Pages Incorporate NRC Requested Changes,Per Recent Telcon1999-05-27027 May 1999
Forwards Revised marked-up TS & UFSAR Pages to 990303 LARs 259 & 131 Which Revised Qualifications for Operations Mgt & Incorporated Generic Position Titles in Ts.Encl Pages Incorporate NRC Requested Changes,Per Recent Telcon
L-99-082, Dockets Licensee Plan for Bvps,Unit 1,safety-related Small Bore Piping Evaluation Project Discussed in NRC 990311 Public Meeting at BVPS1999-05-17017 May 1999
Dockets Licensee Plan for Bvps,Unit 1,safety-related Small Bore Piping Evaluation Project Discussed in NRC 990311 Public Meeting at BVPS
L-99-071, Notifies of License Withdrawal for J Scott,License SOP-11481,due to Resignation from Employment at BVPS1999-05-12012 May 1999
Notifies of License Withdrawal for J Scott,License SOP-11481,due to Resignation from Employment at BVPS
1999-09-03
[Table view]
Category:UTILITY TO NRC
MONTHYEARML20059L4271990-09-0707 September 1990
Requests Approval for Use of Steam Generator Tube Plugs for Both Mechanical & Welded Applications
ML20059G0821990-09-0404 September 1990
Forwards Application for Amend to License DPR-66,consisting of License Change Request 180,changing Section 3.3.3.2 to Reduce Required Number of Operable Incore Detector Thimbles for Remainder of Cycle 8
ML20059F7551990-08-29029 August 1990
Responds to Unresolved Item 50-334/90-16-01 Noted in Insp Rept 50-334/90-16.Corrective Actions:Initial Training for Maint Group Personnel Responsible for Maintaining Supplied Air Respirators Will Be Supplemented W/Biennial Retraining
ML20059F1501990-08-29029 August 1990
Advises That Permanent Replacement Chosen for Plant Independent Safety Evaluation Group.Position Will Be Staffed Effective 900829
ML20028G8731990-08-29029 August 1990
Forwards fitness-for-duty Program Performance Data for Jan-June 1990,per 10CFR26.71
ML20059D3761990-08-24024 August 1990
Describes Cycle 3 Reload Design,Documents Util Review Per 10CFR50.59 & Provides Determination That No Tech Spec Changes or Unreviewed Safety Questions Involved.Reload Core Design Will Not Adversely Affect Safety of Plant
ML20028G8881990-08-24024 August 1990
Withdraws Operator License SOP-10731 (55-60749) Issued to K Gilbert,Who Resigned
05000412/LER-1990-007, Responds to NRC Re Deviation 50-412/90-12-01 Noted in Insp Rept 50-412/90-12.Corrective Actions:Procedure OM 2.20-4.I Revised to Require Removal of Flanges Following Drain Operations & LER 90-007-00 Issued1990-08-23023 August 1990
Responds to NRC Re Deviation 50-412/90-12-01 Noted in Insp Rept 50-412/90-12.Corrective Actions:Procedure OM 2.20-4.I Revised to Require Removal of Flanges Following Drain Operations & LER 90-007-00 Issued
ML20058P7651990-08-14014 August 1990
Provides Info on Acceptability of Rescheduling Response to Reg Guide 1.97 Ser,Item 4b, Neutron Flux Monitoring Instrumentation. Rescheduling of Util Response Will Be Determined on or Shortly After Meeting W/Nrc
ML20059E0571990-08-10010 August 1990
Forwards Suppl 3 to Nonproprietary WCAP-12094 & Proprietary WCAP-12093, Evaluation of Pressurizer Surge Line Transients Exceeding 320 F for Beaver Valley Unit 2, for Review by 900901.Proprietary Rept Withheld (Ref 10CFR2.790(b)(4))
ML20059E7631990-08-0101 August 1990
Provides Results of Util Evaluation of Licensed Operator Requalification Exam Conducted During Wks of 900709 & 16. Crew That Failed to Meet Expected Performance Level Has Been Successfully Upgraded & re-evaluated to Be Satisfactory
ML20059B8141990-08-0101 August 1990
Requests Exemption from 10CFR26 Re Fitness for Duty Program & 10CFR73 Re Physical Protection of Plants & Matls Concerning Unescorted Access Requirements for Nuclear Generating Stations
ML20056A3471990-07-31031 July 1990
Responds to NRC Bulletin 90-001.Items 1 Through 5 of Requested Actions for Operating Reactors Completed
ML20056A1841990-07-27027 July 1990
Forwards Revised Methodology for Achieving Alternate Ac for Plant,Per 900720 Telcon
ML20055H2581990-07-25025 July 1990
Forwards Decommissioning Rept, Per 10CFR50.33(K) & 50.75(b)
ML20055F7061990-07-0909 July 1990
Responds to NRC Re Dcrdr Requirements as Specified in Suppl 1 to NUREG-0737.DCRDR Corrective Actions Implemented & Mods Determined to Be Operational Prior to Startup Following Seventh Plant Refueling Outage
ML20055D3871990-07-0202 July 1990
Provides Info Re long-term Solution to Action Item 3 of NRC Bulletin 88-008,per 890714 & s.Util Will Continue to Monitor Temp in Affected Lines & Evaluate Results
ML20058K5031990-06-29029 June 1990
Discusses Use of Emergency Diesel Generators as Alternate Ac Source at multi-unit Sites,Per Licensee .Emergency Diesel Generator Load Mgt Methodology Evaluated to Meet Listed Criteria
ML20044A3661990-06-21021 June 1990
Forwards Application for Amend to License NPF-73,consisting of Tech Spec Change Request 44,changing Stroke Time to 60 for Inside Containment Letdown Isolation Valves.Change Determined Safe & Involves No Unreviewed Safety Issue
ML20043G6811990-06-14014 June 1990
Forwards Application for Amends to Licenses DPR-66 & NPF-73, Revising Tech Specs Re Electrical Power Sys - Shutdown & Ac & Dc Distribution - Shutdown
ML20043H9341990-06-14014 June 1990
Forwards Issue 1 to Rev 4 to Inservice Testing Program for Pumps & Valves. Issue 1 Removes Relief Requests Requiring Prior NRC Approval & Adds Certain Program Changes Permitted by ASME XI & Generic Ltr 89-04
ML20043G5981990-06-12012 June 1990
Forwards Monthly Operating Repts for May 1990 for Beaver Valley Units 1 & 2 & Revised Rept for Apr 1990 for Beaver Valley Unit 1
ML20043G6851990-06-12012 June 1990
Forwards Application for Amend to License DPR-66,consisting of Proposed OL Change Request 176,revising Tech Specs to Replace Current Single Overpressure Protection Setpoint W/ Curve Based on Temp
ML20043G7941990-06-12012 June 1990
Responds to NRC 900524 Request for Addl Info Re Proposed Operating License Change Request 156.Clarification of Magnitude of Confidence Level of Westinghouse Setpoint Methodology,As Specified in WCAP-11419,encl
ML20043G8001990-06-11011 June 1990
Forwards Application for Amend to License NPF-73,consisting of Proposed Operating License Change Request 41.Amend Deletes Surveillance Requirement 4.4.9.3.1.d
ML20043H0291990-06-11011 June 1990
Forwards Application for Amend to License NPF-73,consisting of Proposed OL Change Request 40,modifying Heatup & Cooldown Curves Applicable to 10 EFPYs Per WCAP-12406 Re Analysis of Capsule U from Radiation Surveillance Program
ML20043F5251990-06-0707 June 1990
Requests Temporary Waiver of Compliance from Tech Spec Limiting Condition for Operation Re Operability of Containment Isolation Valves During Quarterly Slave Relay Testing.Evaluation to Support Request Encl
ML20043F1361990-06-0404 June 1990
Advises That Chemistry Manual Chapter 5P1, Enhanced Primary to Secondary Leakrate Monitoring Program for Unit 1,per 880328 Request to Recommit to Item C.1 of NRC Bulletin 88-002
ML20043B5971990-05-18018 May 1990
Advises of Delay in Hiring Independent Safety Evaluation Group Replacement to Maintain Five Permanent Personnel Onsite,Per Tech Spec 6.2.3.2.Replacement Will Be Provided within 30 Days of Retirement of Engineer on 900531
ML20043B0511990-05-15015 May 1990
Responds to Telcon Request for Addl Info Re Elimination of Snubbers on Primary Component Supports.Probability of Case B/G Event Extremely Small & Does Not Represent Realisitic Scenario
ML20043B1921990-05-11011 May 1990
Forwards Cycle 8 & Cycle 2 Core Operating Limits Rept,Per Tech Spec 6.9.1.14
ML20042G9761990-05-0808 May 1990
Responds to NRC 900226 Ltr Re Violations Noted in Insp Repts 50-334/89-80 & 50-412/89-80.Corrective Action:Maint Work Request Program Being Upgraded to Include Responsibilities of Nuclear Const Dept & Will Be Issued by 900601
ML20042G8541990-05-0303 May 1990
Forwards Technical Review,Audit Summary & Operability Assessments Re Potentially Invalid Leak Detection Tests Used as Alternative for Amse Section XI Hydrostatic Tests
ML20042G9071990-05-0101 May 1990
Forwards Annual Financial Repts for Duquense Light Co,Ohio Edison Co,Pennsylvania Power Co,Centerior Energy Corp & Toledo Edison Co,Per 10CFR50-71(b)
ML20042F1381990-04-30030 April 1990
Advises That Final SER for Implementation of USI A-46 Will Be Delayed Until Late 1990
ML20042F0991990-04-20020 April 1990
Forwards Response to Request for Addl Info Re Second 10 Yr ISI Program
ML20012F5951990-04-10010 April 1990
Forwards Monthly Operating Repts for Mar 1990 & Revised Operating Data Rept & Unit Shutdown & Power Reductions Sheets for Jan 1990
ML20042E1471990-04-0404 April 1990
Forwards Application for Amends to Licenses DPR-66 & NPF-73, Consisting of License Change Request 174/36,updating Staff Titles to Reflect Nuclear Group Organization
ML20012F6021990-03-30030 March 1990
Submits Supplemental Response to Station Blackout Rule for Plant,Per NUMARC 900104 Ltr.Summary of Changes to Condensate Inventory of Dhr,Effects of Loss of Ventilation, Control Room HVAC & Reactor Coolant Inventory Listed
ML20012E3091990-03-23023 March 1990
Forwards Response to 900308 Request for Addl Info on Reg Guide 1.97 Re Variable for Steam Generator wide-range Level Instrumentation
ML20012E3451990-03-23023 March 1990
Submits Addl Info for Exemption from General Design Criteria GDC-57,including Background Info Describing Sys Operation & Addl Bases for Exemption Request.Simplified Recirculation Spray Sys Drawings Encl
ML20012D6491990-03-19019 March 1990
Requests Retroactive NRC Approval of Temporary Waiver of Compliance Re Tech Spec Limiting Condition for Operation 3.8.2.1 on Ac Vital Bus Operability.Sts Will Be Followed When Inverters Not Providing Power to Vital Bus
ML20012E4091990-03-16016 March 1990
Forwards Inservice Insp 90-Day Rept,Beaver Valley Power Station Unit 1,Outage 7, for 880227-891221,per Section XI of ASME Boiler & Pressure Vessel Code 1983 Edition Through Summer 1983 Addenda,Section XI
ML20012D6181990-03-15015 March 1990
Responds to NRC 900215 Ltr Re Violations Noted in Insp Repts 50-334/89-23 & 50-412/89-22.Corrective Actions:Safety Injection Signal Reset & Plant Returned to Presafety Injection Conditions & Crew Members Counseled
ML20042D7401990-03-14014 March 1990
Forwards Corrected Annual Rept of Number of Personnel Receiving Greater than 100 Mrem & Associated Exposure by Work Function at Plant for CY89.
ML20012D5801990-03-13013 March 1990
Forwards Correction to First 10-yr Inservice Insp Program, Rev 2 to Relief Request BV2-C6.10-1 Re Recirculation Spray Pump - Pump Casing Welds & Relief Request Index
ML20012D6221990-03-13013 March 1990
Forwards Response to Generic Ltr 89-19, Resolution to USI A-47. Recommends All Westinghouse Plant Designs Provide Automatic Steam Generator Overfill Protection to Mitigate Main Feedwater Overfeed Events
ML20012C1791990-03-0909 March 1990
Responds to NRC 900207 Ltr Re Deviations Noted in Insp Repts 50-334/89-25 & 50-412/89-23.Corrective Actions:Written Request Initiated to Identify Unit 2 post-accident Monitoring Recorders in Control Room & Recorders Labeled
ML20012E0911990-03-0505 March 1990
Lists Max Primary Property Damage Insurance Coverages for Plant,Per 10CFR50.54(w)(2)
ML20012B7051990-03-0202 March 1990
Responds to NRC 900131 Ltr Re Violations Noted in Insp Repts 50-334/90-05 & 50-412/90-04.Requests Withdrawal of Violation Re Stated Transport Problem & Reclassification as Noncompliance,Per 10CFR2,App C,Section G
1990-09-07
[Table view] |
Text
.
'Af Telephone (412) 393-6000 Nuclear Group P.O. Box 4 Shippingport PA15077-0004 October 10, 1985 ector of Nuclear Reactor Regulation United' States Nuclear Regulatory Commission Attn: Mr. Hugh L. Thompson, Jr. Director Division of Licensing Washington, DC 20555
Reference:
Beaver Valley Power Station, Unit No. 1 Oocket No. 50-334, License No. OPR-66 Implementation of TMI Action Item II.K.3.5, "Autmatic Trip of Reactor Coolant Pumps" (Generic Letter No. 85-12)
Gentlemen:
In response to your letter of June 28, 1985, Generic Letter 85-12, the following information is submitted regarding the Reactor Coolant Pump (RCP) trip issue. In particular, we nave selectec appropriate RCP trip criteria based upon the Westingnouse Owner's Group (WOG) methodology whicn nas been included in the WOG Emergency Response Guidelines (ERGS). In our response of April 18, 1983, to Generic Letter 83-10d, we indicated our plans to include tne WOG methodology in our new symptom-based Emergency Operating Procedures for Beaver Valley Power Station, Unit No. 1 (BVPS-1) and have since utilized the WOG ERGS, Revision 1 for this purpose. The schedule for the implementation of the new E0Ps has since been confirmed by Commission Order dated the 12th of June 1984. ,
In reviewing the WOG RCP trip criteria, the Commission notes that the process of criterion selection involves a number of plant specific considerations. Accordingly, your letter indicated the need to provide the information requested in Section IV of the Safety Evaluation for the WOG (Enclosure to Generic Letter 85-12). The information contained in the Attachment to this submittal details our responses to eacn requested item of Section IV in order to complete our response to Generic Letter 83-10d and subsequently Generic Letter 85-12.
If you have any questions regarding this submittal, please contact myself or memoers of my staff.
Very truly yours,
$ O
$34 Pag ice Pr ident, Nuclear O
n-80 aver Valley Power Station, Unit No. 1 Docket No. 50-334, License No. DPR-66 Implementation of TMI Action Item II.K.3.5, " Automatic
-Trip of Reactor Coolant Pumps" (Generic Letter No. 85-12)
Page 2-cc: Mr. W. M. Troskoski, Resident Inspector U. S. Nuclear Regulatory Commission Beaver Valley Power Station Shippingport, PA 15077 U. S. Nuclear Regulatory Commission c/o Document Management Branch Washington,-DC 20555 Director, Safety Evaluation & Control Virginia Electric & Power Company P.O. Box 26666 One' James River Plaza Richmond, VA 23261 l
1 l
l ATTACHMENT Response to Generic Letter 85-12 Implementation of TMI Action Item II.K.3.5
" Automatic Trip of Reactor Coolant Pumps" l
A DETERMINATION OF RCP TRIP CRITERIA A.1 Identify the instrumentation to be used to determine the RCP trip setpoint, including the degree of redundancy of each parameter signal needed for the criterion chosen.
Response
The following is based upon the WOG methodology as provided in the ERGS, Revision 1, for the calculation of alternative RCP trip criteria, and consideration of . instrumentation qualification and uncertainties. We nave calculated the BVPS-1 RCP trip setpoints for each of the alternative parameters and have selected the Secondary Pressure Dependent.RCS Pressure criterion. With this method, the RCS Pressure Setpoint is determined based on the actual steam generator pressure. We have determined tnat the selected parameters satisfy the discrimination tequirements in NRC Generic Letter 83-10d. Provided below is the requested information for the instrumentation to be used for the criterion chosen.
RCS Wide-Range Pressure (0-3000 PSIG) P-RC-402 and P-RC-403
- a. Senscrs (PT-RC-402, 403)
Rosemount Model 1154GP9RB Gage Transmitter
- b. Racks (P-RC-402, 403) g 7100 Series Rack Modules
- c. Indicators (PI-RC-402A, 403):
y Model 43-VX-252 Steam Generator Pressure (0-1400 PSIG) P-MS-474,475,476 /484,485,486 /494,495,496
- a. Sensors (PT-MS-474,475,476 /484,485,486 /494,495,496)
ITT Barton Model 763 Gage Transmitter
- b. Racks (P-MS-474,475,476 /484,485,486 /494,495,496)
W 7100 Series Rack Modules
- c. Indicators (PI-MS-474, 474A, 475, 476, 484, 485, 485A, 486, 494, 495, 496, 496A)
W Model 43-VX-252 The wide-range RCS pressure indicators are safety grade quality and redundancy exists (2 channels-loop A and C). For the secondary pressure indicators, redundancy exits for each loop (3 channels per loop). The above instruments are environmentally qualified for the design basis accident under which they are required for mitigation.
Att chment
. Page 2 A.2(a) Identify the instrumentation uncertainties for both normal and adverse containment conditions.
Response
Normal Indication Accuracy (NIA)*
Loops: P-RC-402/403 NIA = 1 71. PSIG Loops: P-MS-474,5,6 / 484,5,6 / 494,5,6 NIA = 1 40 PSIG Post Accident Indicator Accuracy (PIA)
- Loops: P-RC-402/403 PIA = 1 371 PSIG Loops: P-MS-474,5,6 / 484,5,6 / 494,5,6 PIA = 1 250 PSIG
- Instrument uncertainty values are rounded upward to the nearest full PSI for additional conservatism.
A.2(b) Describe the basis for selection of the adverse containment carameters:
Response
'The ERGS, Revision 1, have been used to develop the BVPS-1 new E0Ps.
The parameter setpoints (i.e. RCP Trip Parameters, etc.) in the E0Ps have been established by using the WOG methodology and by considering the associated instrument uncertainties. One factor which affects the instrument uncertainties is the environmental conditions which vary for
'the type and severity of an accident. Two sets of parameter values for the RCP trip criterion are provided in the E0Ps, one for normal conditions and one for post accident conditions, i.e. adverse containment conditions. The operator uses one or the other set of values, depending on containment pressure and radiation conditions. The selection .of the pressure and radiation values which determine adverse
-containment conditions is based on the ERG defined values (
Reference:
ERG, Executive Volume, Generic Instrumentation Section). If either the containment pressure exceeds 5 psig or the containment radiation exceeds 10E5 R/hr., the ERG values, tne E0P adverse containment (post accident) parameter setpoint is implemented by the operator. If the containment conditions are less than these values, the E0P normal parameter setpoint is-used.
A.2(c) Address, as appropriate, local conditions such as fluid jets or pipe whip whicn might influence the instrumentation reliability.
Response
A review of plant drawings and instrument locations, including sensing lines and electronics, has been completed and it has been determined
. Attr.chment-
. Page 3 that no single failure as described in our UFSAR will result in a fluid jet or pipe whip resulting in the common cause loss of all instrumentations used to monitor the RCP trip setpoint. Sufficient redundancy and separation exists to preclude this from occurring.
A.3 In addressing the selection of the criterion, consideration to uncertainties associated with the WOG supplied analyses values must be provided. These uncertainties include both uncertainties in the computer program results and uncertainties resulting from plant specific features not representative of the generic data group.
Response
The LOFTRAN computer code was used to perform the alternate RCP trip criteria analyses. Both Steam Generator Tube Rupture (SGTR) and non-LOCA event were simulated in these analyses. Results from the SGTR analyses were used to obtain all but three of the trip parameters.
LOFTRAN is a Westinghouse licensed code used for FSAR, SGTR and non-LOCA analyses. The code has been validated against the January 1982 SGTR event at the Ginna plant. The results of this validation show that LOFTRAN can accurately predict RCS pressure, RCS temperatures and
. secondary pressures especially in the first ten minutes of the transient. This is the critical time period when minimum pressure and subcooling is determined.
The major causes of uncertainties and conservatism in tne computer program results, assuming no changes in the initial plant conditions (i.e. full power, presurizer level, all SI and AFW pumps running) are due to either models or inputs to LOFTRAN. The following are considered to have the most impact on the determination of the RCP trip criteria:
- 1. Break flow
- 2. SI flow
- 3. Decay heat
- 4. Auxiliary feedwater flow The following sections provide an evaluation of the uncertainties associated with each of these items.
To conservatively simulate a double ended tube rupture in safety analyses, the break flow model used in LOFTRAN includes a substantial amount of conservatism (i.e., predicts higher break flow than actually expected). Westinghouse has performed analyses and developed a more realistic break flow model that has been validated against the Ginna SGTR data. The break flow model used in tne WOG analyses has been shown to be 'approximately 30% conservative when the effect of the higher predicted break flow is compared to the more realistic model. The consequence of the higher predicted break flow is a lower than expected predicted minimum pressure.
The SI flow inputs used were derived from best estimate calculations, assuming all SI trains operating. An evaluation of the calculational methodology shows that these inputs have a maximum uncertainty of + 10%.
n_ ,
Attachment
..T
'Pige14-
!The decay heat'mo' del used in the WOG analyses was based on the 1971 ANS 5.1 standard. When compared -with the-more_recent 1979 ANS 5.1 decay 1 heat : inputs, -the . values used in the WOG analyses are higher by about-
- 5%., To idetermine the effect of the' uncertainty due to the decay heat model, a sensitivity study wasiconducted for SGTR.. The results of this
, study' show :that a: .20% ' decrease in decay heat resulted in only a 1%
decrease Ein ;RCS pressure for the first 10 minutes of the transient.
Since JRCS temperature.is controlled by steam relief, it is not affected by the decay heat model uncertainty- .
.The AFW flow. rate input used .in the WOG analyses are best estimate values, : assuming that all auxiliary feed pumps are running, minimum pump start delay, and no throttling. ETo evaluate the uncertainties with AFW flow,' rate, a sensitivity study was performed. Results from the two loop plant _ study show that,:a 64% increase in AFW flow'resulted in only an 8%
. decrease in minimum. RCS pressure, a 3% decrease in minimum RCS
- subcooling, and an 8% decrease in minimum pressure _ differential.
-Results from the 3 loop plant study show that, a 27% increase in AFW
-flow resulted in only ,a 3% decrease in minimum RCS pressure, a 2%
idecrease_'in minimum RCS subcool'ng, and a 2% decrease in pressure differential,
~The effects of all~ these- uncertainties with the models and input parameters were evaluated and it was concluded that the contributions from the break flow conservatism and the SI uncertainty dominate. The calculated overall uncertainty in the WOG analyses as a result of these considerations for the Beaver Valley Power Station Unit 1-is -30 to L+300 PSI for the RCS/ Secondary Differential Pressure RCP trip setpoint.
Due to the minimal effects from the decay heat model and AFW input, these results' include only the effects of the uncertainties due to the creak-flow model and SI flow inputs.
-8 POTENTIAL REACTOR COOLANT PUMP PROBLEMS
.B.1 Assure that containment isolation, including inaavertent isolation will not cause problems if it occurs for non-LOCA transients and accidents.
- a. Demonstrate that, if water services needed for RCP operations are terminated, they can be restared fast enough once a non-LOCA situation is confirmed to prevent seal damage or
. . failure.
- b. Confirm that containment isolation with continued pump
-operation will not lead to seal or pump damage or failure.
Response-The. water services needed for RCP operation includes a) the reactor plant component cooling water system (CCR) which supplies cooling water for the 'RCP thermal barrier neat exchangers and for the RCP motor bearings oil neat exchangers and b) the RCP seal injection flow which is provided by the charging /HHSI pumps.
. Attachment
.. P g] 5 l l
l The- design and operation of the RCP seal injection flow is such that it
. remains in service under anticipated plant transients. Neither containment isolation phase A or B (CIA or CIB) will isolate RCP seal injection flow. The seal injection flow by itself is adequate for providing RCP seal cooling which prevents seal damage or failure.
Therefore, RCP seal damage or failure is precluded due to the continued
. supply of seal injection water.
l The CCR flow is maintained during a CIA signal which therefore permits continued RCP operation. In the event a CIB occurs, CCR flow to the RCP's is terminated which will result in stopping the RCPs within L
approximately five minutes. CCR flow is necessary for RCP operation since it is the source of cooling for the RCP motor bearings oil heat exchangers, however, limited operation is possible without pump damage or failure occurring. CCR flow cannot be restored fast enough following a CIB signal to maintain the RCP's running. The new symptom-based Emergency Operating Procedures (EOPs) provide restart instructions in the event RCP operation is desired. This is consistent with the ERG
-background documents and the general philosophy that RCPs should be operating to aid in plant cool down and RCS pre;,sure control for non-LOCA transients and accidents.
Additional information on containment isolation is provided in UFSAR Section 5.3.3. Table 5.3-1 identifies the various containment isolation arrangements and states what isolation signals exist for these water systems supporting RCP operation.
The following provides assurance that upon receipt of a CIB signal, the
-operator will take the appropriate action.
- 1. Through training, the operators are instructed that if CCR water flow to the RCPs is isolated on a containment pressure signal, all the RCPs should be stopped within 5 minutes because of loss of RCP motor bearing oil cooling. The current E0Ps contain cautions to this effect where appropriate.
- 2. In the new symptom-based E0Ps, one of the operator immediate action steps in E-0, " Reactor Trip or Safety Injection", instructs the operator to stop all RCPs upon verification that CIB is required. A subsequent E-0 step instructs the operator to stop all RCPs if there is no CCR flow to the RCPs.
- 3. The operator would be alerted to the RCP conditions and the need to trip the RCPs since one or several monitors would alarm in the control room. These alarms are symptomatic of the loss of CCR water flow and include the following:
RCP Alarms Related to CCR Flow Loss:
- 1. RCP Lower Bearing Lube Oil Cooling Water Flow Low
- 2. RCP Stator Winding Cooling Water Flow Low I
Ja l-
, ! Attachment'
. 'Page 6
- 3. RCP Upper Bearing Lube Oil Cooling Water Flow Low
- 4. RCP Motor Bearing Temperature High
- 5. RCP Lower. Radial Bearing Temperature High
- 6. RCP Cooling Water Discharge Temperature High
- 7. RCP Thermal Barrier Cooling Water Discharge Temperature High Alarms Related to the CCR System
- 1. CCR Water Surge Tank Level Hign-Low
- 2. CCR Water Heat Exchanger 8" Discharge Line Flow Low-
- 3. CCR Water Heat Excnanger 14" Discharge Line Flow Low
- 4. CCR Pump Discharge Pressure Low
- 5. CCR Pump Auto Start-Stop.
- 6. High Temperature Alarms on Numerous Components.
r-It is notea that, if RCP operation (at least one pump) becomes desirable, -the new E0Ps provide for RCP restart unaer the appropriate conditions (see response to C2 herein for RCP Restart steps). Also, there are two function restoration procedures, FR-C.1 and FR-P.1, where special RCP restart (i.e., without RCP support conditions required) is considered necessary. The conditions for RCP restart are discussed in the ERGS, Executive Volume, RCP Trip / Restart, Section 2.4 and provided in the symptom-based E0Ps.
B.2 Identify. the components required to trip the RCPs, including relays, power supplies and breakers. Assure that RCP trip, when determined necessary, will occur.- If necessary, include the effects of adverse containment conditions on RCP trip reliability. Describe the basis for the adverse _ containment parameters selected.
Response
Based on manual RCP trip, the following set of components and locations are identified for each RCP. The mark numbers or other identification
.information are delineated below for RCP-1A, 18 and 10, respectively.
Component Description Identification
- 1. Air Circuit Breakers'(ACB) 1AS, 185, and ICS
- 2. Local Control Power Breakers 125VOC Heineman
- 3. Cell Swite.hes 52H-1A5, 52H-185, & 52H-1C5 1
o Attachment
,_ P:ge 7 4 Auxiliary Contacts 52-1A5,52-185, & 521C5
- 5. Disconnecting Contacts. One each for eacn ACB
- 6. Trip Coils One each for each ACB
- 7. Control Power Supp'y (125VDC) Bus 1-1A Breaker 1, Bus 1-1A Breakers Breaker 2, and Bus 1-2A Breaker 1.
- 8. Control-Switches 1-1AS, 1-185, and 1-1C5 Components 1 through 6 are locatea in the normal switchgear room, components 7 in the process instrument room, and components 8 are located on the bench board in the control room. It is not necessary to 1
-include the affects of adverse containment conditions on RCP trip reliability since the above equipment is not subject to high energy line breaks and the environment of the rooms in which they are located is not anticipated to be significantly different than the environment that would occur during normal-plant operation.
Assurance that RCP trip, when determined necessary, will occur is providea by the following:
- 1. Preventive maintenance having a 18 month frequency is performed on the RCP breakers. - '
- 2. The RCPs are controlled from the oenchboards in the control room.
Switch positions are STOP -
START with red (running) and white (normal-shutdown or bright-auto trip) status indicating lights.
Also available in the control room, are motor ammeter indication for each RCP and flow indication for each reactor coolant loop.
In the event that this instrumentation snows that an RCP has not tripped when manually tripped from the main control board by the operator, the operator can then manually trip the RCP at the pump breaker in the switchgear room.
C OPERATOR TRAINING AND PROCEDURES (RCP Trip)
C.1 Describe the operator training program for RCP trip. Include the general philosophy regarding the need to trip pumps versus the desire to keep pumps running.
Response
The selection and implementation of the RCP trip criterion is based on the WOG methodology. The methocology nas been chosen within the WOG generated Emergency Operating Procedures (EOPs) and is being presented to all licensed operators and STAS during the Training Program for these new E0Ps. A description of this pnase of training and philosophy is stated below.
RCP operation is of a concern during small break LOCA events. Accident analysis determinations are based on the assumption that offsite power is lost at the onset of the l
3 JAtt:chmentL
. :Page.8 event. This is . not necessarily a conservative assumption during a. small break .LOCA. .The peak cladding temperature s t limit of ,2200*F may be exceeded if, the RCP's trip after reaching; a- critical time -period. The RCP trip criteria is based -on tripping the RCP's prior to reaching this critical time.
If RCP's are not running during the event, RCS inventory will decrease to a level below .the break elevation. When.this occurs only. steam is released out of the break. With RCP's running, a steam water mixture is forced out of the break.
This .results .in a' larger amount of' mass loss.from the RCS.
The forced . circulation, however, also results in an increase in core- cooling. The critical time period is when the
-increased inventory' loss results .in a deeper core uncovery when the pumps trip and.the core has not been cooled'enough to prevent exceeding the peak cladding temperature limit.
The RCP trip criteria and. training ensures that the operator will trip the pumps prior to reaching the critical time period. The increase in RCS mass loss due to RCP's running does not occur until the break would have uncovered had pumps
'not been running. The break cannot uncover until the primary si_de of- the _S.G. U-tubes have drained and this cannot begin until -saturation . conditions are reached at the top of the-U-tubes, thereby allowing steam to fill the void created by draining. The saturation conditions at the top of the U-tubes depends on- RCS pressure and temperature and the conditions in the secondary system. The RCS temperature is
' dependent on. the-secondary temperature and its corresponding secondary pressure. The secondary pressure may vary due to the means of = steam relief being used. For example, the secondary pressure vill be less if steam is dumped to the main . condenser rat ar than relieved through -the code safeties. For the actual secondary pressure and the associated : primary to secondary . temperature differential needed for~ heat transfer through the U-tubes, there is a corresponding RCS pressure which indicates that the RCS is approaching saturation at the top of the U-tubes. The resultant crimary to secondary pressure differential provides the criterla for tripping the RCPs.
The criterion for tripping the RCP's is RCS pressure 145 psi greater than the highest SG pressure for normal containment conditions and 510 psi for adverse containment conditions.
This number is derived from summing primary to secondary, delta-P. corresponding to the delta-T required for heat transfer, height difference of the U-tubes to the pressure instruments, steamline delta-P, and instrument inaccuracies.
If primary to secondary delta-P decreases to the trip setpoint of 145 psi and high head SI flow is verified, the RCP's must be tripped.
. .m ., , _ - _. .. - _ _ . _ . - _ _ _ . . . . . . _ _. . __. .- .
- Attachment-P:ge 9 L4
, - - j0uring. non-LOCA . events, ,it is desired to keep RCP's running -
to provide. normal RCS pressure. control, (i.e., spray flow) p and . vessel- head cooling. If pumps are tripped during a:SGTR
! j.c .or 'other non-LOCA events, pressurizer PORV's would have to be utilized to reduce RCS pressure. -The preferrea method is to .
utilize pressurizer sprays which are most effective when the RCPs -are operating. The 145 psi primary to secondary delta-P trip criteria ~.provides protection ~ during .small break LOCA
- events, while ensuring pumps are not. tripped during a design
- basis: SGTR- event of a~ single tube,,and selected non-LOCA
- events. The non-LOCA events analyzed are a steam break and a
-.feedline -break. The analysis shows that primary:to secondary-delta-P should not decrease below 350 psi for these events.
. RCP.'s will not be tripped during a SGTR or non-LOCA event H ~ unless_ containment . pressure rises- to 10 psig and CIB ;
actuates. The resulting loss of CCR to-the RCP's would then require that they be tripped.
C.2. Identify those procedures which inclu'de RCP trip related operations.
- a. RCP trip using WOG alternate criteria
'b. RCP restart ,
l
- c. Decay. heat removal by natural circulation
- d. Primary sytem void removal e .- Use of steam generators with and without RCPs operating, f-. RCP trip for other reasons.
Response
The BVPS-1 .EOPs have been developed based on the WOG ERGS, Revision 1,
~
l
.and the E0P -designations are the same as in the ERGS. 'We believe that-
'this response and the review of this response -can therefore -be
,_ .faciliated by reference to Table 4, Section 2.6, of the ERG Generic i Issue, .RCP Trip / Restart and by identifying any differences for the
.BVPS-1 ~E0Ps. The ERG Table 4 provides a summary of RCP Trip and Restart- "
steps as contained in the BVPS-1 E0Ps except as noted below..
g ECA-3.2 . Add a special trip step (ST) relative to inadequate RCS subcooling and RVLIS not operable.
FR-C.1 Add a special restart without support conditions required.
(SR)
FR-P.1 Replace the Table 4 restart criteria (R) with SR (Refer-ence Section 2.4.4, ERG Generic Issue RCP Trip / Restart). ;
E-0, E-1, Add an "or" condition in conjunction with each trip (T).
E-3 and-
~
ECA-2.1 L ECA-1,1 .. Add an "or" condition in conjunction with trip (TI). The l , conditions for these last two groups of trips relate to loss [
of reactor plant component cooling water (CCR) flow.
l
- , , ,, _.-....,---,.__..-.___,_.m.._-~,--,- ,- . , - . - - - . , , _ . .
~
f'" w
-, :Attacnment-Pag 3 10 LE Add ST for RCPs supplying failed spray valve.
Future -changes to these E0Ps would be performed in accordance with our program as. described in our Procedures Generation Package submitted for NRC review on June 28,- 1984.
4
--